1. Field of the Invention
The present invention relates to zirconium alloys, particularly, to zirconium alloys for use in fuel cladding and structural applications within nuclear reactor vessels, and, more particularly, to zirconium alloys having improved corrosion resistance in aggressive water chemistry environments during the operation of boiling water reactors (BWR) and may have some utility in pressurized water reactors (PWR).
2. Background Art
Nuclear reactors are used in electric power generation, research and propulsion. A reactor pressure vessel contains the reactor coolant, i.e., water, which removes heat from the nuclear core. Piping circuits are used to carry the heated water or steam from the pressure vessel to the steam generators or turbines and to return or supply circulated water or feedwater to the pressure vessel. Typical operating pressures and temperatures for the reactor pressure vessels can be about 7 MPa and 288° C. for BWRs and about 15 MPa and 320° C. for PWRs. The materials used in these respective environments must, in turn, be formulated and/or manufactured to withstand various loading, environmental (high-temperature water, oxidizing species, radicals, etc.) and radiation conditions to which they will be subjected during extended operation of the reactor.
BWR and PWR typically include nuclear fuel sealed in cladding comprising one or more layers of metal or metal alloys to isolate the nuclear fuel from the moderator/coolant system, i.e., water in PWRs and steam and/or water in BWRs. The cladding typically includes at least one layer of a zirconium-based alloy including one or more alloying element and include layers of both a zirconium alloy and unalloyed zirconium. Cladding may also utilize a composite system having an inner lining of sponge zirconium or dilute zirconium alloy containing minor amounts, less than about 0.5 wt % of iron or other elements, as alloying metals. Typically, the cladding will be configured as a tube in which pellets of the nuclear fuel are stacked to fill substantially the entire length of the cladding tube. The tubes will then be arranged in bundles, with a plurality of bundles being arranged to define the reactor core.
Under normal operating conditions, zirconium-based alloys are useful as a nuclear fuel cladding material due to their relatively low neutron absorption cross sections and, at temperatures below about 398° C., their strength, ductility, stability, and lack of reactivity in the presence of demineralized water or steam. “Zircaloys” are a widely used family of commercially-available, corrosion-resistant, zirconium-based alloy cladding materials that include 97-99% by weight zirconium, with the balance being a mixture of tin, iron, chromium, nickel and oxygen. Two particular alloy compositions, specifically Zircaloy-2 and Zircaloy-4, are widely used for manufacturing cladding although Zircaloy-2 is the more commonly utilized composition for BWR applications.
In addition to zirconium, Zircaloy-2 includes about 1.2-1.7 wt % Sn; 0.07-0.20 wt % Fe; 0.05-0.15 wt % Cr, and 0.03-0.08 wt % Ni. Zircaloy-4, on the other hand, although including similar quantities of the other alloying elements present in Zircaloy-2, is substantially free of nickel and has an Fe concentration of about 0.18-0.24 wt %.
The presence of these alloying elements, which are relatively insoluble in zirconium under normal conditions, will generally result in the formation of Second Phase Particle (SPP) “precipitates” in an α-phase zirconium matrix. Under equilibrium conditions, the alloy matrix will be a single phase with the alloying elements present at concentrations at or near their respective solubility limits. The formation of precipitates results from the presence of alloying elements in concentrations above their solubility limits. For example, the precipitates most commonly found in Zircaloys may be generally represented by the chemical formulas Zr(Fe,Cr)2 and Zr2(Fe,Ni).
Cladding corrosion occurs in both BWRs and PWRs with the corrosion typically occurring in nodular or uniform forms. Corrosion in nodular form is generally more prevalent in BWRs. Nodular corrosion is usually a porous near-stoichiometric zirconium oxide forming on the surface of the cladding. It can rapidly cover the entire surface of the Zircaloys in small, localized patches (referred to as “nodules” or “pustules”) with thinner uniform corrosion in between. Uniform corrosion tends to be more prevalent in PWRs, and typically consists of a uniform layer of zirconium oxide forming on the surface of the cladding. The uniform layer typically contains a small excess of zirconium, appears as a black or gray film and exhibits semiconductive properties.
Normally the degree of uniform or nodular corrosion is acceptable and does not limit nuclear reactor operations. In some low frequency abnormal circumstances the degree of corrosion can become excessive and lead to through-wall cladding penetration and thus release highly radioactive species to the coolant and limit reactor operation.
Some corrosion failure mechanisms are now understood well enough to limit their occurrence. One such mechanism that occurs in BWRs is known as Crud Induced Localized Corrosion (“CILC”). The CILC mechanism involves a combination of cladding susceptible to nodular corrosion and a high concentration of copper in the coolant. The primary source of copper is from corrosion dissolution of brass materials used in steam condenser construction. Copper infiltrates the nodular oxide layer and creates a localized region that has low thermal conductivity thus leading to localized overheating and accelerated corrosion.
The problem of CILC has been addressed by controlling the coolant purity and minimizing cladding nodular corrosion. To control the coolant purity, steam condensers have been replaced with non-copper bearing materials, filtering systems optimized for copper removal are available, and monitoring for copper levels has been established. To minimize cladding nodular corrosion, processes that produce a fine SPP size (i.e., use of β or α+β heat treatments followed by low thermal input to prevent Ostwald ripening) have been implemented and preferable elemental compositions within the ASTM Zircaloy specification have been defined.
As will be appreciated, corrosion control and prevention is extremely important for the safe operation of nuclear reactors and corrosion-induced component failures have the potential for causing serious injury, reactor downtime and reduced efficiency. The physical, chemical and electrochemical interactions between the reactor components and the aqueous environment to which they are exposed during reactor operations are, understandably, significant factors for understanding and controlling corrosion. Accordingly, both the composition and surface conditioning of the reactor components and the composition and purity of the coolant water must be considered and an appropriate combination utilized to provide improved corrosion control.
Indeed, unacceptable levels of corrosion have been attributed to the presence of aggressive water chemistry conditions and its deleterious effect on fuel cladding materials. It is also believed that temporary excursions from the preferred reactor operating conditions can result in greatly accelerated corrosion rates. Thus, although the fuel cladding utilized in a reactor may have been processed in accord with the best practices recognized in the prior art for controlling corrosion, the use of such materials in aggressive water chemistry conditions and/or its exposure to periodic excursions may result in unacceptable corrosion rates, thereby increasing the risk of corrosion failures and the maintenance cost. The prior art knowledge includes alloy compositions within the ASTM specification for Zircaloy-2 as well as other Zr-based alloys such as that described in U.S. Pat. No. 4,664,727, a late stage solution heat treatment as outlined by U.S. Pat. Nos. 4,450,016, 4,576,654, and 5,437,747, restricted thermal input subsequent to the solution heat treatment as outlined by Japanese Patent Publications No. 3172731/2001. Despite the knowledge and development efforts represented by these prior art references, corrosion and the risk of corrosion failures is a continuing problem in the nuclear industry that past experience, design specifications and controls have not been able to eliminate completely. Further improvements toward preventing or suppressing corrosion remain necessary to achieve the goal of 100% fuel reliability desired for improved nuclear reactor operation and reduced maintenance costs, particularly in reactor systems that are, or may be, exposed to aggressive water chemistry conditions, whether the result of intentional addition of water conditioning packages, local conditions and/or episodic excursions from the desired water chemistry. As a result, there remains a need for improved cladding materials that can increase the operating margin of a reactor system by providing improved resistance to aggressive water chemistry environments.
Unfortunately, the particular chemistry and/or the particular condition that produce an aggressive water condition within the reactor water environment is often not well characterized, particularly in the event of excursions from standard operating conditions, such that variations in Zircaloy corrosion performance can occur between BWRs that operate with similar nominal reactor water chemistry. Transient aggressive environments, where one or more chemical species, known or unknown, are inadvertently introduced to the reactor coolant over a short period of time are, by their nature, difficult to detect and quantify. Robust cladding that can tolerate aggressive water chemistry environments without incurring unacceptable corrosion rates and increased risk of failure is highly desirable.
Impurities may be unintentionally introduced into the reactor water by various means such as spilling of cutting, cleaning, or hydraulic fluids, leaking of steam condenser tubes that carry impure secondary cooling water, incomplete cleaning following piping chemical decontamination operations, and/or compromised filtering equipment. Impurities from such sources may be in such low concentration that they go undetected and yet may still trigger accelerated cladding corrosion.
The corrosion kinetics of zirconium alloys typically exhibit two stages, for alloys such as the Zircaloys that contain SPPs of relatively insoluble transition metals such as Fe, Cr, Ni, V, etc. The initial corrosion typically comprises the diffusion-limited growth of a thin oxide film on the metal surfaces. Once this oxide film exceeds a thickness of about 2 μm, the film formation can begin to break down and may transition to an approximately linear growth phase with the multiple stages of diffusion-limited growth and breakdown occurring over extended exposure periods.
Previous approaches for controlling corrosion have included various modifications to the concentrations of alloying elements (particularly iron and nickel) in Zircaloy alloys to reduce the severity of nodular corrosion by increasing the availability of aliovalent ions that, in turn, improves the uniformity of the oxide.
The SPPs play an important role in the corrosion behavior of the alloy(s) in which they are formed with the precipitate composition, average precipitate size and the precipitate distribution (i.e., the interparticle spacing) affecting, perhaps significantly, the corrosion properties of the particular alloy. An approach commonly pursued in parallel with alloy chemistry control involves controlling the size and distribution of the SPPs, particularly within the surface regions of the reactor fuel assembly components. As a result of the difference in corrosion mechanisms acting in BWRs and PWRs, conventional Zircaloy cladding compositions are prepared differently with those intended for use in PWR applications being subjected to higher temperature anneals and slow quenches (less than 5° C./second) to produce relatively larger precipitate sizing. Conversely, cladding compositions intended for use in BWR applications utilizing lower temperature anneals and with fast quenches (greater than 5° C./second and more typically greater than 20° C./second) to produce relatively smaller precipitate sizing.
To improve the intergranular and intragranular distribution of SPPs within Zr—Sn—Fe alloys of the present invention, the alloy may be heated into the β-phase temperature range, e.g., above about 1000° C., to form a solid solution that is substantially free of SPPs. The β-phase alloy can then be rapidly quenched to produce a substantially diffusionless martensitic transformation, particularly in the surface regions exposed to the quenching composition. By cooling the alloy rapidly, i.e., a rate greater than about 500° C./second, through the α+β-phase temperature range, approximately 825-965° C., and into the α-phase range, typically below about 800° C., the alloying elements will tend to remain in a supersaturated metastable solution in the zirconium matrix. At slower cooling rates, however, the alloying elements will tend to nucleate and grow SPPs whose final size depends on the cooling rate, with slower quench rates resulting in relatively larger SPPs. Subsequent heat treatments in the α-phase after the rapid quench will allow Zr(Fe, Cr)2 and Zr2(Fe, Ni) SPPs to grow, or to nucleate and grow from the metastable solid solution. Although the size and distribution of SPPs can be controlled to some extent by thermal-mechanical processing, in order to prevent excessive growth of the SPPs it is necessary to limit the subsequent thermal exposure of a Zircaloy component after an initial heat treatment to dissolve the SPPs.
Accordingly, zirconium alloys are now extensively used as fuel cladding materials and fuel assembly materials in BWRs, PWRs and other nuclear applications. As noted above, two of the most common zirconium alloys in use are Zircaloy-2 and Zircaloy-4. Additional details regarding the specific alloys corresponding to Zircaloy-2 and Zircaloy-4, are provided in U.S. Pat. Nos. 2,772,964 and 3,148,055, the disclosures of which are hereby incorporated by reference in their entirety.
In addition to the basic composition of the alloy, conventional techniques for reducing or preventing nodular corrosion include heat treatment methods in which an alloy is heated, for a short period of time, to a temperature at which the alloy exists in α+β or β phase, after which the alloy is rapidly quenched, to control the microstructure. Such a process is described in Japanese Patent Publications Nos. 45699/1986 and 58223/1988 and the application of such a method in connection with a particular alloy composition is detailed in Japanese Laid-Open Patent Publications Nos. 43450/1985 and 228442/1987. Similarly, another approach for providing improved nodular corrosion resistance involves applying a heat treatment to only the outer region of the cladding tube as detailed in U.S. Pat. No. 4,576,654.
In order to continue improving the cladding performance and reactor efficiency, there continues to be a need to develop zirconium alloys that have increased corrosion resistance in adverse water chemistry conditions and can be manufactured efficiently and economically.